ML20246K927
| ML20246K927 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/30/1989 |
| From: | William Cahill, Walker R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEIN-87-019, TXX-89644, NUDOCS 8909060172 | |
| Download: ML20246K927 (9) | |
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Log # TXX-89644 l
L File # 836
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Ref. # 10CFR50.36 1tIELECTRIC August 30, 1989' William J. CahlII, Jr.
Executiae M President
'i U. S. Nuclear Regulatory Commission
]
Attn: Document Control Desk d
Washington, DC 20555 L
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
I DOCKET NOS. 50-445 AND 50-446 o
CPSES UNIT 1 TECHNICAL SPECIFICATIONS
. Gentlemen:
All Westinghouse nuclear power pressurized-water reactor (PWR) facilities holding an operating. license or construction permit were not fied (IE Information Notice No. 87-19) of the potential for perforation and cracking of rod cluster control assemblies (RCCAs).
Subsequently, accelerated RCCA wear was obr,erved in several Westinghouse plants with 17x17 fuel and the " upper head f-cold" reactor upper internals design.
In sone instances, accelerated weJ2 produced through-wall clad penetration after only one cycle of operation. The accelerated fretting wear on the RCCAs has been observed at locations on the rodlets corresponding to the control rod drive guide card positions when the RCCAs are in their fully withdrawn position. Since CPSES utilizes 17x17 fuel and the " upper head T-cold" reactor upper internals, a plant-specific change to the Technical Specifications is proposed to mitigate the effects of accelerated RCCA wear.
Attached are proposed changes to the CPSES Unit 1 Technical Specifications which will mitigate the effects of accelerated RCCA wear at CPSES by permitting the implementation of Westinghouse recommended wear mitigation techniques. The proposed changes permit the repositioning of the RCCA "all-ods-out" por! tion within the interval of 222 through 231 steps during operation which will result in even distribution of the wear along the rodlet surface.
-Similar changes to Technical Specifications have been approved and are on the dockets of the Wolf Creek and Callaway facilities. Westinghouse has assessed the impact of the RCCA repositioning guidelines on CPSES and has identified no significant safety issues.
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8909060172 890830 PDR. ADOCK 05000445 Q
FDC i
400 Nonh Olive Street LB81 Dallas, Texas 75201
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.. _], z TXX-89644 August 30, 1989
'l Page 2~of 2 i
The following is a list of attachments providing the above described information:
l Attachment A:
Description of the changes to the CPSES' Unit l' Technical l
Specifications'which.will allow the-implementation of Westinghouse recommended wear. mitigation techniques.
Attachment B:
Marked-up pages of the certified CPSES Unit 1 Final Draft Technical. Specifications.
Attachmera C:
Typed version of'the. proposed changes to the certified CPSES l
Unit 1 Technical Specifications.
I TU Electric requests that the NRC perform'an expedited review of the above changes'and inform us as to their. acceptability, j
Sincerely.
Y
' -d s
William J. Cahill, Jr.
By:
tm 41 b-Roger [j. Walker l
Manager, Nuclear Licensing j
JTB/vid j
i c - Mr. R. D. Martin, Region IV l
Resident Inspectors. CPSES (3) j i
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CPSES ' VS AM3NE]}NT '.1 DE1 AILED DESCRIPTION' Page 1' Page 1 of 1 4
y
- TSi Page (as amended)
Group Description 3/4 1-19. 20-2
.See Page No(s):22 Implementation of Westinghouse recommended control rod wear mitigation techniques.
Revision:
IE.Information Notice No. 87-19 notified Westinghouse facilities of potential' for perforation and cracking-of RCCAs. The proposed changes to the. Technical-Specifications allows-the implementation of recommended ~ wear mitigation techniques.
Similar changes have been. accepted by the'NRC for Wolf Creek &
Callaway facilities. Westinghouse'has assessed-the impact of the change for CPSES and has' concluded that the shutdown margin will not be violated and that the MTC values are bounded by values assumed in the safety analyses.
Additionally the effect on Fz and Axial Offset is at most +0.5% relative to the all-rods-out calculation (when the RCCAs are parked furthest in the core).
TS - Change Request Number: TS-89-097 SER/SSER Impact: No
- 0 I
1
hb 59b Page 1 of 3 REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION-f0R OPERATION h sie D rs. Ce 3.1.3.4 The individual (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.4 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
T,yg greater than or equal to 551'F, and a.
b.
All reactor coolant pumps operating.
APPLICABILITY:
MODES 1 and 2.
ACTION:
With the drop time of any rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of rods shall be demonstrated through measurement prior to reactor criticality:
a.
For all rods following each removal of the reactor vessel head, b.
For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and c.
At least once per 18 months.
I COMANCHE PEAK - UNIT I 3/4 1-19
'3 f
REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION TOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn.
) pap _f APPLICABILITY: MODES 1* and 2" **.
ACTION:
With a maximum of one shutdown rod not fully withdrawn, Except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour either:
a.
Fully withdraw the rod,.or b.
Declare the rod to be inoperable and apply Specification 3.1.3.1.
SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:
a.
Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and b.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
l l
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
- With K greater than or equal to 1.
t Foll itLJ n 3}xil bene cc J:Curj dert si,u+dovo ceJ> ott Cl 9
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k 222 umN
- 23) Skep u llbrm 90$dlon a 4hein}truulO f
t COMANCHE PEAK - UNIT I 3/4 1-20
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--w v
(27. 3,22t)
[81.6,222) 240.
22 20-t
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200 h
BANK B
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#I 160 3
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. ER BANK C
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120 0
b m
80 YZ BANK D
<O (0.49)
C (31.0)
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20 40 60 80 100 PERCENT OF RATED THERMAL POWER A Yell w Odrown 360 kMe cendTNm ubm cadaloh on4 I y
apos# ton w:% 4b h4Mul of ?zit cad S zJI shp ud{),w,,
FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER COMANCHE PEAK - UNIT 1 3/4 1-22
Att'achment:C-to TXX-89644
' August!31',1989i
~
} Page.1' of 3 REACTIVITY' CONTROL SYSTEMS-1' ROD DROP TIME 1'
LIMITING CONDITION FOR OPERATION 3.1.3.4.The individual-(shutdown and control) rod drop time from the DRAFT physical fully withdrawn. position shall be less than or equal to 2.4 -
seconds from beginning of decay of stationary gripper coil voltage to dashpot. entry with:
Tavg greater than or equal to 5510F, and a.
b.
All reactor coolant pumps operating.
APPLICABILITY: MODES I and 2.
ACTION:
With the drop time of any rod determined to exceed the above
. limit, restore the rod drop time to within the above limit prior to proceeding to MODES 1 or 2.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The_ rod drop time of rods shall be demonstrated through measurement prior to reactor criticality:
a.
For all rods following each removal of the reactor vessel head,.
b.
For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific l
rods, and c.
At least once per 18 months.
I 1
COMANCHE PEAK - UNIT 1 3/4 1-19 DRAFT l-l
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' Attachment C to TXX-89644-August 31, 1989
' Page 2 of 3 -
E REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn.+
DRAFT APPLICABILITY: MODES 1* and 2*#.
ACTION:
With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour either:
a.
Fully withdraw the rod, or b.
Declare the rod to be inoperable and apply Specification 3.1.3.1.
SURVEILLANCE RE0UIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:
a.
Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and b.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
- With Kerr greater than or equal to 1.
+ Fully withdrawn shall be the condition where shutdown rods are at a position within the internal of 1222 and $231 steps withdrawn DRAFT COMANCHE PEAK - UNIT 1 3/4 1-20 DRAFT
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Attachment C to TXX-89644 August 31':1989 Page 3 of 33 240
,f231..
j (81.6,222)
(27.3,222)
{222 I-2 200 3
BANK B 1
EO f
- (0,164)
- 160 3
g (100,146)7 BANK C
>-8 120 T-Z l
9t-m 80 N
BANK D
/
/(0,49) 4E Q
40
-~
O m
, (31,0) 0 20 40 60 80 100 PERCENT OF RATED THERMAL POWER
- Fully withdrawn shall be the condition where control rods are at a position witMn the interval of >_222 and $231 steps withdrawn.
Figure 3.1-1 ROD BAhX INSERTION LIMITS Vi.RSUS THERMAL POWER Comanche Peak - Unit 1-3/4 1-22
(